1. Field of the Invention
This invention relates to nuclear reactors which circulate cooling water by natural circulation power, and to nuclear reactors which circulate at most 30% of the cooling water by forced-circulation power.
This invention improves critical power performance at the time of a pressure rise transient (rather than critical power performance at the time of a water-supply temperature transient) by making a dynamic void coefficient greater than a predetermined value.
2. Description of the Related Art
A conventional fuel assembly and a conventional reactor core are explained with reference to FIG. 13 and FIG. 14. FIG. 13 is a sectional view showing a principal part inside a conventional reactor core, and shows the arrangement of a fuel assembly and a control rod. In a fuel assembly 1, two water-rods 5 where coolant (i.e. non-boiling water) flow inside are arranged and, for example, two or more fuel rods 3 containing nuclear fuel material 6 (like a cylinder pellet) are enclosed as shown in this figure. A channel box 4 surrounds the fuel assembly 1, and a boiling water stream path 8 is formed in the channel box 4.
The fuel assembly 1 and a control rod 2 are arranged regularly, in the reactor core 14. A control rod 2, for example, is arranged four per fuel assembly 1.
The gap between a fuel assembly and an adjacent assembly serves as a by-pass portion 7 in which the coolant (i.e. non-boiling water) exists. The cooling water in the by-pass portion 7 does not boil, even if there is heating by neutron radiation, and the like.
The parameters d1 and d2 in FIG. 13 represent the width of a channel box 4, and a width of the by-pass portion 7, respectively. The width d2 of the conventional by-pass portion 7 is less than 10% of the width d1 of the channel box 4. In the conventional fuel assembly, the water rod(s) 5 is arranged at the central section of the fuel assembly 1.
The following method is adopted in regard to the enrichment distribution of the fuel. The enrichment distribution is arranged in the vertical direction. The enrichment of the upper part is set 0.2 wt % (or less) larger than that of the lower part. There may be installed a natural uranium blanket of a low reactivity at a vertical edges of the fuel. In this case, the upper and lower blankets are the same length, or the upper blanket is longer than the lower blanket.
FIG. 14 is a system figure showing the outline of the conventional boiling water reactor. As shown in this figure, the nuclear reactor core 14 is built in the pressure vessel 11 of the boiling water reactor. The upper part of the nuclear reactor core 14 is surrounded with a shroud 13. A steam separator 12 to separate steam and saturated water is installed in the upper part of the shroud 13. By the steam separator 12, the steam is drawn up and the saturated water is led to the outside of the shroud 13.
Underneath the nuclear reactor core 14, a plurality of control rods 15 are installed. The control rods 15 are driven in and out with control rod drive 16 installed in the lower part of the pressure vessel 11.
Usually, the internal pressure of the pressure vessel 11 during operating power is set as 70 times atmospheric pressure, for example. The shroud 13 divides coolant flow to the upper part of the pressure vessel from the nuclear reactor core 14 (the arrow with attached symbol 10a in the figure) and flow exterior of the nuclear reactor core 14, i.e., the coolant flow to the lower portion of the pressure vessel produced near the inner wall of the pressure vessel 11 (the arrow with attached symbol 10b in the figure).
Based on differences in the mechanisms which cause the flow of such coolant, there are two kinds of boiling water reactors, that is, a natural-circulation reactor and a forced-circulation reactor. In the natural-circulation reactor, cooling water is driven by natural-circulation power by the saturated water outside of the shroud 13. Thus, the saturated water is led inside to the lower part of the nuclear reactor core 14.
In contrast, the saturated water circulates in a forced-circulation water reactor by a drive from a power apparatus installed outside of the shroud 13. As the power apparatus, there are employed a re-circulating water pump, an internal pump, etc.
A main steam pipe 17 is connected to the pressure vessel 11, and the steam generated by the nuclear reactor is led to a high-pressure turbine 20. A plurality of relief safety valves 18 are set in the main steam pipe 17. When an abnormal pressure rise happens, a relief safety valve 18 is opened and the internal pressure of the pressure vessel 11 is reduced.
A turbine governor valve 19 is set between the high-pressure turbine 20 and the relief safety valves 18 to adjust the amount of steam introduced to the high-pressure turbine 20.
When a so-called pressure rise transient phenomena occurs, such as during a loss of a generator load, in order to prevent the rotational frequency of the turbine from going too high, the turbine governor valve 19 is closed. When the turbine governor valve 19 is closed, the main steam is usually led to a condenser 23 through a by-pass line 28.
A low-pressure turbine 21 is installed downstream of the high-pressure turbine 20, and rotation of the turbines is converted into current by a generator 22 installed downstream of the low-pressure turbine 21. The steam working in the turbine passes through a steam extraction line 29a from the low-pressure turbine 21, is led to the condenser 23, and is liquefied.
Through a feed pipe 27 and a feed pump 26, the cooling water (i.e. condensation) made by the liquefying is returned to the pressure vessel 11, and circulates through it. In the feed pipe 27, a low-pressure feed water heater 24 and a high-pressure feed water heater 25 are provided. These feed water heater 24 and 25 heat the condensation to suitable water-supply temperature conditions. Feed water heater 24 and 25 operate by taking the steam extraction from each stage as a heat source to heat the coolant water (i.e. condensation) to the appropriate temperature condition. That is, heating of the condensation is performed by the steam extraction from the low-pressure turbine 21 and the high-pressure turbine 20 through steam extraction lines 29b and 29c, respectively, in the low-pressure feed water heater 24 and the high-pressure feed water heater 25. The cooling water temperature of the outlet of the high-pressure feed water heater 25 is about 70 subcool temperature.
The boiling water reactor ensures a stability margin while operating under the following circumstances, by considering beforehand the worst conditions, such as a change in (xcex94MCPR) of the minimum critical power ratio (MCPR):
1) the internal pressure of the pressure vessel 11 experiences an abnormal rise;
2) an unusual transient change during operating power occurs in the form of a water supply temperature change outside of the normal range.
MCPR represents the minimum value of the ratio between fuel assembly power (critical power) expected when the boiling transition begins to happen and actual output. When the boiling transition begins to occur, since a liquid layer covering the fuel rod surface will be lost, the cooling state of the fuel rod surface will get worse, and the fuel temperature will rise.
Pressure rise transients occur due to, for example, loss of a load, and water-supply temperature transients occur due to improper operation of a temperature control unit and the like.
In the forced-circulation water reactor, although a flow rate transient due to failure of forced-circulation equipment is assumed, such a failure does not exist in a natural-circulation water reactor.
In connection with the pressure rise transient, the design assumes introduction of the steam to the by-pass line is successful. Generally, since the phenomena at failure is worst case, a stability margin is secured by basing the design on these failure conditions.
In the case of the pressure rise transient of a forced-circulation water reactor, shut down of recirculation is performed at least in part, the power of the nuclear reactor is controlled, and as a result xcex94MCPR is kept small. For example, the method of stopping (xe2x80x9cshut downxe2x80x9d) one internal pump (or several sets) may be adopted at the time of a pressure rise transient in a forced-circulation water reactor in which ten internal pumps are installed.
However, in a natural-circulation water reactor, this type of shutdown (by controlling pumps) cannot be accomplished when an abnormal pressure rise transient during operating power occurs, compared with the forced-circulation water reactor mentioned above.
Therefore, in natural-circulation reactors, there is the characteristic that xcex94MCPR during a pressure rise transient becomes larger than in the forced-circulation water reactor. Moreover, compared with the forced-circulation water reactor, it is difficult to stop the circulating flow rate of the coolant in a natural-circulation nuclear reactor. If a circulating flow rate equivalent to a forced-circulation water reactor is provided, a big head difference is needed. Therefore, the height of the pressure vessel will need to be increased as compared to the forced-circulation water reactor. When the height of the pressure vessel of the natural-circulation water reactor is made to be about the same as the forced-circulation water reactor average, the following problems will arise.
Since the coolant circulating flow rate is decreased as compared with the forced-circulation water reactor, the average void fraction of the reactor core becomes larger. The dynamic void coefficient also increases, compared with the dynamic void coefficient of a forced-circulation water reactor.
Therefore, negative reactivity feedback at the time of a pressure rise transient does not function effectively, and xcex94MCPR becomes larger after all.
In a nuclear reactor, that permits boiling water in the reactor core, the amount of steam bubbles in the reactor core changes with a change of thermal power, pressure, coolant flow, water-supply temperature, and the like. The void coefficient (also called void reactivity coefficient) represents the ratio of change of reactivity accompanied by a change of the amount of steam bubbles in the reactor core, and is an important factor in the safety and the stability of the nuclear reactor.
Thus, a conventional natural-circulation nuclear reactor has the characteristic that xcex94MCPR during a pressure rise transient tends to become large, as compared to the forced-circulation water reactor. Therefore, the stability margin needs to be increased.
In the conventional natural-circulation water reactor, the power density is limited to provide adequate core stability. In other words, if pressure rise transient properties and in-core stability are controlled, a smaller reactor core power density, compared with the forced-circulation water reactor, results.
This makes a natural-circulation reactor less efficient and more costly than a forced-circulation reactor.
The invention is made in order to solve the problems discussed above.
The invention provides an acceptable void coefficient under various conditions and at the same time controls the increase in volume. In-core stability is improved, power density is increased, and an economical natural-circulation reactor (or partial forced-circulation reactor) is achieved.
The invention provides a reactor core, cooled primarily by natural-circulation, with a void reactivity coefficient between xe2x88x920.07 and xe2x88x920.03% xcex94k/k/% void fraction. This void reactivity coefficient range is achieved by, for example, the design of the by-pass portion and channel box, the enrichment distribution along the axial direction, the provision of blanket areas, and/or the arrangement of water rods and fuel-rods-within a channel box.